Yuanzineng kexue jishu (Jan 2024)
TEM Examination of M5 Zirconium Alloy Cladding of Spent Fuel Rod
Abstract
The zirconium alloy cladding of PWR fuel rods which undergo high neutron irradiation during service, will cause significant changes in its microstructure, thereby affecting its macroscopic performance. Therefore, the study of neutron irradiation behavior of zirconium alloy cladding is a focus of nuclear field. However, due to the strong radioactivity of materials after neutron irradiation, relevant experiments must be conducted in a hot cell. Therefore, research on the microstructure of irradiated fuel cladding is a difficult task. In this study, the microstructure of M5TM zirconium alloy cladding material after neutron irradiation was studied by means of transmission electron microscope in the hot cell facility of China Institute of Atomic Energy. The samples were from commercial pressurized water reactor AFA3G type spent fuel rods with burnup of 14 GW·d/tU and 41 GW·d/tU, respectively. A cladding sample with a length of about 10 mm from the fuel rod was cut, and the defueling and chemical cleaning in the hot cell were carried out to obtain a clean cladding sample. Then, mechanical sampling methods was used to prepare a thin slice sample of the cladding with 3 mm diameter. Finally, the electrolytic twin-jet thinning method was used to prepare the cladding transmission electron microscopy observation and analysis sample. In addition, to compare the structural changes of zirconium alloy cladding before and after irradiation, the same method was used to prepare un-irradiated observation and analysis samples of the same material. The observation and analysis results of the un-irradiated and irradiated samples reveal that there are native second phase particles (SPPs) inside the matrix structure of the un-irradiated zirconium alloy cladding, and the overall interior of the matrix is with few nano precipitates and no obvious dislocation structure observed. After irradiation, there is no significant difference in the size and distribution of the native SPPs in the matrix compared to the un-irradiated sample, but significant nano precipitates and high-density dislocation structures appear. As the fuel burnup increases, the size of nano precipitates increases. The similarity of dislocation structures between low and high burnup samples indicates that under the burnup of 14 GW·d/tU, the dislocation structures generated by irradiation in the zirconium alloy cladding basically reach saturation state. The results of selected area electron diffraction (SAED) indicate that although there are some amorphous structures in the native SPPs in the matrix after irradiation, the bcc crystal structure is still the main structure, indicating that the SPPs maintain certain irradiation stability at the burnup of 41 GW·d/tU. In addition, the EDS results of the SPPs indicate that with the increase of fuel burnup, the content of Nb element tends to be depleted. Analysis suggests that after neutron irradiation, the Nb atoms in the SPPs of zirconium alloy expand into the Zr matrix, promoting the precipitation of Nb elements in the form of nano Nb rich phases in the Zr matrix.
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